1 RESULTS ON MODELING OF PRIMARY WATER STRESS CORROSION CRACKING AT CONTROL ROD DRIVE MECHANISM NOZZLES OF PRESSURIZED WATER REACTORS Omar Fernandes Aly 1 , Arnaldo H. Paes de Andrade 2 , Miguel Mattar Neto 3 , Mnica Schvartzman 4 ABSTRACT One of the main failure mechanisms that cause risks to pressurized water reactors (PWR) is the primary water stress corrosion cracking (PWSCC) occurring in alloys like the alloy 600 (75Ni-15Cr-9Fe). It can occur, besides another places, at the control rod drive mechanism (CRDM) nozzles. It is caused by the joint effect of tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. These cracks can cause problems that reduce nuclear safety by blocking the displacement of the control rods and may cause leakage of primary water that requires repair or replacement of the reactor pressure vessel head. In this work it is performed a study of the existing models and proposed a new approach to assess the primary water stress corrosion cracking in nickel-based Alloy 600 CRDM nozzles . The proposed model is obtained from the superposition of electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained from CDTN-Brazilian Nuclear Technology Development Center, in a SSRT equipment, according with Schvartzman et al.(2005). Staehle (1992) has built a diagram that indicates a thermodynamic condition for the occurrence of some PWSCC submodes in Alloy 600: it was used potential x pH diagrams (Pourbaix diagrams), for Nickel in high temperature primary water (300 0 C till 350 0 C). The PWSCC submodes were located over it, using experimental data. Also, a third parameter called stress corrosion strength fraction was added. However, it is possible to superimpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. It is important to mention that the main contribution of this work is from a specific experimental condition of potencial versus pH, it was superposed, an empiric-comparative, according with Staehle (1992), a semi-empirical- probabilistic according with Gorman et al. (1994), an initiation time according with Garud (1997), and a strain rate damage according with Boursier et al.(1995)-models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. The results were compared with the literature and it showed to be coherent. From this work was obtained a modeling methodology from experimental data. The SSRT tests had been realized at a condition of potential =621 mV SHE and pH= 7.3. The PWSCC strength fraction evaluated was 0.95: this initiates an empirical-comparative model. The initiation time model obtained was according Eq. (1) with t i in days, T in K, and σ in MPa. The model was planned for constant load, but some assumptions were done to obtain (1) from slow strain rate tests. t i = 4,88. 10 -23 . exp (32822, 35/T). ln [1,79 (278,5/σ)] (1) The strain rate damage model obtained was in according to Eq. (2) with t i in days, ė SSRT in s -1 for primary water temperature 303 0 C. t i = 8,28. 10 -3 . ė SSRT -0,67 (2) The semi-empirical-probabilistic model was obtained only for its deterministic part since there wasnt enough number of tests for modeling the probabilistic part. Notwithstanding some assumptions can be done over it and compared with literature. 1 Doctor in Nuclear Technology-Materials , Post- Doctoring, CCTM-IPEN, Sªo Paulo University, Brazil ( [email protected] ) 2 Professor, Doctor in Materials Engineering, CCTM-IPEN, Sªo Paulo University, Brazil ( [email protected] ) 3 Professor, Doctor in Structural Engineering , CEN-IPEN, Sªo Paulo University, Brazil ( [email protected] ) 4 Doctor in Materials Science, Researcher CDTN/CNEN-MG, Brazil ( [email protected] ) SMiRT 19, Toronto, August 2007 Transactions, Paper # G04/2
RESULTS ON MODELING OF PRIMARY WATER STRESS CORROSION CRACKING AT CONTROL ROD DRIVE MECHANISM NOZZLES OF PRESSURIZED WATER REACTORS
Jul 01, 2023
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